ORAU Team

Dose Reconstruction Project for NIOSH

Paducah Gaseous Diffusion Plant – Occupational External 

Dose

Document Number:

ORAUT-TKBS-0019-6

Effective Date: 03/29/2005

Revision No.: 01

Controlled Copy No.:_______

Page 1 of 25

Subject Expert: James E. Turner

Document Owner

Approval: Signature on File Date: 03/22/2005

Jay J. Maisler, TBD Team Leader

Approval: Signature on File Date: 03/22/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 03/22/2005

Richard E. Toohey, Project Director

Approval: Signature on File Date: 03/29/2005

James W. Neton, Associate Director for Science

Supersedes:

Revision No.: 00

TABLE OF CONTENTS

Section

Record of Issue/Revisions......................................................................................................................

Page

3

Acronyms and Abbreviations..................................................................................................................4

6.1 Introduction ....................................................................................................................................5

6.2 Basis of Comparison......................................................................................................................5

6.3 Dose Reconstruction Parameters..................................................................................................6

6.3.1 Administrative Practices........................................................................................................6

6.3.2 Dosimetry Technology ..........................................................................................................9

6.3.2.1 Beta/Photon Dosimeters ..............................................................................................9

6.3.2.2 Neutron Dosimeters .....................................................................................................9

6.3.3 Calibration...........................................................................................................................10

6.3.3.1 Beta/Photon Dosimeters ............................................................................................10

6.3.3.2 Neutron Dosimeters ...................................................................................................11

6.3.4 Workplace Radiation Fields ................................................................................................11

6.3.4.1 Beta/Photon Fields.....................................................................................................11

6.3.4.2 Workplace Beta/Photon Dosimeter Response...........................................................13

6.3.4.3 Neutron Fields............................................................................................................13

6.3.4.4 Workplace Neutron Dosimeter Response..................................................................14

6.4 Adjustments to Recorded Dose ...................................................................................................14

6.4.1 Photon Dose .......................................................................................................................14

6.4.2 Nonpenetrating Dose ..........................................................................................................14

6.4.3 Neutron Dose......................................................................................................................14
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6.5 Missed Dose ................................................................................................................................14

6.5.1 Estimating Missed and Unmonitored Photon Deep Dose...................................................15

6.5.2 Estimating Missed and Unmonitored Shallow Dose ...........................................................16

6.5.3 Estimating Missed Neutron Dose........................................................................................19

6.6 Uncertainty...................................................................................................................................19

6.7 Dose Reconstruction ...................................................................................................................19

6.8 Organ Dose .................................................................................................................................20

References ...........................................................................................................................................21

Glossary................................................................................................................................................23

LIST OF TABLES

Table

6-1 Dosimeter type, period of use, exchange frequency, MDL and potential 

annual missed dose....................................................................................................................

Page

8

6-2 Major radiation sources.............................................................................................................12

6-3 Buildings and periods of operation............................................................................................13

6-4 Technetium-99 recovery operations .........................................................................................13

6-5 Average recorded deep dose and maximum for any single worker by year.............................17

6-6 Average recorded shallow dose and maximum for any worker by year ...................................18

6-7 Shallow dose equivalent rates for 99Tc .....................................................................................19

LIST OF FIGURES

Figure

6-1 Estimated dosimeter photon response characteristics .............................................................

Page

10

6-2 Historical distribution of deep dose equivalent..........................................................................16

6-3 Historical distribution of shallow dose equivalent......................................................................18
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 3 of 25

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 12/16/2003 00-A New technical basis document for the Paducah 

Gaseous Diffusion Plant – Occupational External 

Dose. Initiated by Jay J. Maisler.

Draft 01/08/2004 00-B Incorporates internal review comments. Initiated 

by Jay J. Maisler.

Draft 01/23/2004 00-C Incorporates NIOSH and additional internal review 

comments. Initiated by Jay J. Maisler.

Draft 05/05/2004 00-D Incorporates additional input from Gaseous

Diffusion Plant task team comments. Initiated by 

Jay J. Maisler.

Draft 06/29/2004 00-E Incorporates additional NIOSH review comments. 

Initiated by Jay J. Maisler.

08/24/2004 08/24/2004 00 First approved issue. Initiated by Jay J. Maisler

Draft 11/16/2004 01-A Incorporates results of additional research and 

responses to outstanding comments. Initiated by 

Jay J. Maisler.

Draft 12/10/2004 01-B Incorporates responses to comments from formal 

internal review. Initiated by Jay J. Maisler.

Draft 03/14/2005 01-C Incorporates responses to comments from NIOSH

review. Initiated by Jay J. Maisler.

03/29/2005 03/29/2005 01 Approved issue. Initiated by Jay J. Maisler.
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 4 of 25

ACRONYMS AND ABBREVIATIONS

cm centimeter

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act

GM geometric mean

GSD geometric standard deviation

Hp(d) personal dose equivalent at tissue depth d (d = 10 mm or 0.07 mm)

hr hour

IARC International Agency for Research on Cancer

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1 thousand electron volts

MDL minimum detection level

MED Manhattan Engineering District (a DOE predecessor agency)

MeV megaelectron-volt, 1 million electron volts

mg milligram

min minute

mm millimeter

mrem millirem

NIOSH National Institute for Occupational Safety and Health

NTA nuclear track emulsion, type A (film)

ORNL Oak Ridge National Laboratory

PGDP Paducah Gaseous Diffusion Plant

RU recycled uranium

TLND thermoluminescent neutron dosimeter

TEPC tissue-equivalent proportional counter

TLD thermoluminescent dosimeter

U.S.C. United States Code

yr year
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6.1 INTRODUCTION

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning the preparation of dose reconstructions at particular sites or categories of sites. 

They will be revised in the event additional relevant information is obtained about the affected site(s). 

These documents may be used to assist the National Institute for Occupational Safety and Health 

(NIOSH) in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy facility” as defined in the Energy Employees

Occupational Illness Compensation Program Act of 2000 (EEOICPA) [42 U.S.C. Sections 7384l (5) 

and (12)].

Paducah Gaseous Diffusion Plant (PGDP) workers, especially those employed during the peak 

production decades (1950s, 1960s, and 1970s), have been exposed to radiation types and energies 

associated with enrichment of natural and recycled uranium (RU). PGDP used facility and individual 

worker monitoring methods to measure and control radiation exposure to workers. Before about July 

1960, personnel dosimeters were not assigned to all workers. Records of radiation dose to those 

individuals who wore dosimeters are available beginning in 1953. Doses from these dosimeters were 

recorded at the time of measurement, routinely reviewed by PGDP operations and radiation safety 

staff for compliance with radiation control limits, and routinely available to individual workers. The 

NIOSH External Dose Reconstruction Implementation Guideline (NIOSH 2002) indicates that these 

records represent the highest quality record for assessment and reconstruction of doses. 

Initial radiation dosimetry practices were based on experience gained during several decades of 

radium and X-ray medical diagnostic and therapy applications. These practices were generally well 

advanced at the start of the Manhattan Engineering District (MED) program to develop nuclear 

weapons, beginning in about 1940. 

6.2 BASIS OF COMPARISON 

Since the start of the MED in the early 1940s, various radiation dose concepts and quantities have 

been used to measure and record occupational dose. The basis of comparison for reconstruction of 

dose is the personal dose equivalent, Hp(d), where d identifies the depth (in millimeters) and 

represents the point of reference for dose in tissue. For weakly penetrating radiation of significance to 

skin dose, d = 0.07 mm and is noted as Hp(0.07). For penetrating radiation of significance to wholebody dose, d = 10 mm and is noted as Hp(10). Both Hp(0.07) and Hp(10) are the radiation quantities 

recommended for use as the operational quantity for radiological protection by the International 

Commission on Radiation Units and Measurements (ICRU 1993). In addition, Hp(0.07) and Hp(10)

are the radiation quantities used in the U.S. Department of Energy (DOE) Laboratory Accreditation 

Program (DOELAP) used to accredit the Department’s personnel dosimetry systems since the 1980s 

(DOE 1986). The International Agency for Research on Cancer (IARC) Three-Country Combined 

Study (Fix et al. 1997) and the IARC Collaborative Study (Thierry-Chef et al. 2002) selected Hp(10) as 

the quantity to assess error in historical recorded whole-body dose for workers in IARC nuclear 

worker epidemiologic studies. This technical basis document uses Hp(10) and Hp(0.07) as deep 

dose and shallow dose, respectively.
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 6 of 25

6.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of beta, photon (X- and gamma rays), and neutron energies and geometries of 

exposure, and of the characteristics of PGDP dosimeter responses, are crucial for assessment of the 

original recorded doses. Bias and uncertainty for current dosimetry systems are typically well 

documented (Martin Marietta 1994). The performance of current dosimeters can often be compared 

to the performance of dosimetry systems in the same, or highly similar, facilities or workplaces. In 

addition, current performance testing techniques can be applied to earlier dosimetry systems to 

achieve a consistent evaluation of all dosimetry systems. Dosimeter response characteristics for 

radiation types and energies in the workplace are crucial to the overall analysis of error in recorded 

dose.

Overall, accuracy and precision of the original recorded individual worker doses and their 

comparability to be considered in using NIOSH (2002) guidelines depend on the following factors (Fix 

et al. 1997):

• Administrative practices adopted by facilities to calculate and record personnel dose based 

on technical, administrative, and statutory compliance considerations

• Dosimetry technology, including physical capabilities of the dosimetry system, such as the 

response to different types and energies of radiation, in particular in mixed radiation fields

• Calibration of the respective monitoring systems and similarity of methods of calibration to 

sources of exposure in the workplace

• Workplace radiation fields that could include mixed types of radiation, variations in exposure 

geometries, and environmental conditions

The accuracy of PGDP worker doses has been the subject of DOE investigations (PACE and 

University of Utah 2000). An evaluation of the original recorded doses as available, combined with 

detailed examinations of workplace radiation fields, is the recommended option to provide the best 

estimate of Hp(0.07) for the shallow dose and Hp(10) for the deep dose for individual workers. 

6.3.1

PGDP had a radiation monitoring program using portable instruments, contamination surveys, zone 

controls, and personnel dosimeters to measure exposure in the workplace. The program improved as 

better technology and more information became available. Results from the personnel dosimeters 

were used to measure and record doses from external radiation exposure to PGDP workers. These 

dosimeters include one or more of the following:

Administrative Practices

• Personnel whole-body beta/photon dosimeters 

• Pocket ionization chamber dosimeters

• Personnel neutron dosimeters

For low-energy beta radiation, the dosimeters were likely incapable of furnishing accurate doses in 

terms of Hp(0.07). Extremity doses, which were generally not assessed (PACE and University of 

Utah 2000), were not treated in this analysis.

In 1953, PGDP began using dosimeter and processing technical support provided by the Oak Ridge 

National Laboratory (ORNL). There is evidence that Paducah might have processed its own 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 7 of 25

dosimeters for a period; however, a review of the limited documentation available indicated that 

practices were similar to those used at ORNL and other major sites at that time. Table 6-1 

summarizes PGDP personnel beta/photon and neutron dosimeter characteristics [dosimeter type, 

exchange, minimum detection level (MDL), and potential missed annual dose]. ORNL, which was 

then the Clinton Laboratory, had based its dosimetry methods on the personnel beta/photon 

dosimeter design developed at the Metallurgical Laboratory at the University of Chicago (Pardue, 

Goldstein, and Wollan 1944). ORNL has provided PGDP with dosimeters from early in the operations 

period through the present. 

The precise detection levels listed in Table 6-1 are difficult to estimate, particularly for older systems. 

Current PGDP commercial thermoluminescent dosimeter (TLD) system MDLs are identified in ORNL 

documentation (Martin Marietta 1994) based on a DOELAP-accredited laboratory testing protocol 

(DOE 1986). During earlier years, MDLs were subject to additional uncertainty because factors 

involving radiation field and film type, as well as processing, developing, and reading systems, cannot 

now be tested (Thornton, Davis, and Gupton 1961). The estimates of the film dosimeter MDLs in 

Table 6-1 were based on information from NIOSH (1993), NRC (1989), Wilson et al. (1990), and site 

personnel. Examination of older records, where available, indicated that the Hp(0.07) MDL values 

were about 3 times those for Hp(10) for film. The current TLD MDLs were obtained from ORNL 

(Martin Marietta 1994). 

Parameters of the PGDP administrative practices significant to dose reconstruction involve policies:

• To assign dosimeters to workers

• To exchange dosimeters

• To record notional dose (i.e., some identified value for lower dosed workers, often based on a 

small fraction of the regulatory limit) 

• To estimate dose for missing or damaged dosimeters

• To replace destroyed or missing records

• To evaluate and record dose for incidents

• To obtain and record occupational dose to workers for other employer exposure

PGDP policies appear to have been in place for all these parameters. From startup until July 1960, 

PGDP issued dosimeters to a limited number of individuals (PACE and University of Utah 2000). This 

population of monitored individuals represents those with the highest exposure potential. After July 

1960, PGDP routine practices required the assignment of dosimeters to all workers who entered a 

controlled radiation area (BJC 2000). Dosimeters were exchanged on a routine schedule. For 

workers in some areas the frequency was monthly, but for the general population it was quarterly. All 

dosimeters were processed, and measured results were recorded and used to estimate dose. 

Current administrative practices are generally available (Martin Marietta 1994), as is detailed 

information for each worker in the PGDP exposure history documentation. Summary documents 

provide information on historic practices at PGDP (PACE and University of Utah 2000; BJC 2000).
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Table 6-1. Dosimeter type, period of use, exchange frequency, MDL, and potential annual missed dose.

Dosimeter Period of use Monitored population

Exchange 

frequency

Laboratory 

MDL 

(rem)(a)

Maximum annual 

missed dose 

equivalent (rem)(b)

Hp(10) beta/photon dosimeters

Four-element film dosimeter 1953-7/1960 Selected workers based on activities performed Weekly (n = 50) 0.04 1.0

Four-element film dosimeter After 7/1960 through 1980 Workers in C-340, C-400, and C-410 Monthly (n=12) 0.04 0.24

After 7/1960 through 1980 Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n=4) 0.04 0.08 

After 7/1960 through 1980 Workers and visitors not likely to exceed 0.1 of 

applicable guidelines

Annual (n=1) 0.04 0.02

Harshaw two-chip TLD Beginning 1980 through 1988 Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n=4) 0.02 0.04

Harshaw two-chip TLD Beginning 1980 through 1988 Workers and visitors not likely to exceed 0.1 of 

applicable guidelines

Annual (n=1) 0.02 0.01

Harshaw four-chip TLD, 

8800 series

Beginning 1989 through present Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n=4) 0.02 0.04

Hp(0.07) beta/photon dosimeters

Four-element film dosimeter 1953-7/1960 Selected workers based on activities performed Weekly (n = 50) 0.12 3.0

Four-element film dosimeter After 7/1960 through 1980 Workers in C-340, C-400, and C-410 Monthly (n=12) 0.12 0.72

After 7/1960 through 1980 Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n=4) 0.12 0.24

After 7/1960 through 1980 Workers and visitors not likely to exceed 0.1 of 

applicable guidelines

Annual (n=1) 0.12 0.06

Harshaw two-chip TLD Beginning 1980 through 1988 Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n=4) 0.03 0.06

Harshaw two-chip TLD Beginning 1980 through 1988 Workers and visitors not likely to exceed 0.1 of 

applicable guidelines

Annual (n=1) 0.03 0.015

Harshaw four-chip TLD, 

8800 series

Beginning 1989 through present Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n=4) 0.02 0.04

Neutron dosimetersc

Harshaw TLND Beginning 1998 to 2003 (ongoing) Selected workers based on activities performed Quarterly (n=4) 0.015 0.03

a. Estimated film dosimeter detection levels based on NIOSH (1993), NRC (1989), and Wilson et al. (1990). TLD detection levels from Martin Marietta (1994) and personal communication with site 

personnel. 

b. Maximum annual missed dose (NIOSH 2002).

c. The potential annual missed dose based on laboratory irradiations is not applicable to workplace missed neutron dose. 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 9 of 25

6.3.2

PGDP dosimetry methods evolved as improved technology was developed and complex radiation 

fields were better understood. The adequacy of dosimetry methods to measure radiation dose 

accurately is determined from radiation type, energy, exposure geometry, and other factors described 

in this section. The dosimeter exchange frequency gradually lengthened, corresponding in general to 

the period of regulatory dose controls. 

Dosimetry Technology

6.3.2.1 Beta/Photon Dosimeters

PGDP has historically used personnel dosimeter services from ORNL. In 1945, ORNL implemented 

the beta/gamma film dosimeter design, which was developed originally at the Metallurgical Laboratory 

at the University of Chicago (Pardue, Goldstein, and Wollan 1944). ORNL followed a research and 

development process that led to gradual upgrades in dosimetry capabilities for complex radiation 

fields (Thornton, Davis, and Gupton 1961). Other DOE sites followed this evolution in dosimetry 

capabilities, which led to site-specific multielement film and thermoluminescent dosimetry systems. 

Figure 6-1 shows the energy response characteristics of the PGDP beta/gamma dosimeters based on 

the essentially identical two-element film dosimeter designed at the University of Chicago and used at 

the Hanford Site (as well as ORNL, Los Alamos National Laboratory, and probably other MED sites). 

In addition, Figure 6-1 shows the Hp(10) response. Further, the figure shows the energy response of 

Hanford multielement film and TLDs (Wilson et al. 1990). The curve labeled “Two-Element Film 

Shield” is representative of ORNL dosimeters from 1945 through 1978. ORNL used a multielement 

film dosimeter after 1953 (Thornton, Davis, and Gupton 1961), but processed photon response as it 

did for the two-element dosimeter and used the same shielding as the two-element dosimeter. The 

figure shows that the two-element dosimeter over-responded in relation to Hp(10) from 0.05 to 

0.3 MeV, followed Hp(10) for higher energies, and under-responded for lower energies. Last, the 

figure shows that TLDs are capable of following Hp(10) over the energy range of interest. The 

majority of PGDP worker photon dose comes from handling uranium of low enrichment. The photon 

energy spectrum is almost entirely in the range from 30 keV to 250 keV.

The nonpenetrating response of the two-element dosimeter was calculated as the difference between 

the unshielded and shielded portions of the film based on a uranium calibration. The two-element 

dosimeter workplace nonpenetrating (i.e., beta or shallow) dose response based on the uranium 

calibration should adequately represent Hp(0.07) or at least be claimant-favorable because of the 

significant over-response of the unshielded portion of the film to any lower energy photons that could 

have been present. The multielement film dosimeters and TLDs, which were also calibrated to 

uranium slabs, had the ability to correct more accurately for mixed photon and beta radiation.

6.3.2.2 Neutron Dosimeters

Dosimeters used at PGDP historically had a neutron-sensitive element that was processed on 

request. After 1989, this capability has been provided with a TLD that contained a 6

LiF chip, which is 

very responsive to low-energy neutrons. There is no indication of recorded neutron doses for PGDP 

workers wearing either of these dosimeters. The use of the commercial Harshaw thermoluminescent 

neutron dosimeters (TLND) to assess neutron dose routinely (along with deep and shallow dose) 

began in 1998. ORNL has provided the dosimeter and associated services. The dosimeter has been 

worn with a belt to minimize distance from the worker’s body, which optimizes the albedo effect for 

which the dosimeter is calibrated.
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0.1

1

10

100

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Relative Response

Photon Energy, MeV

Hp(10)

Two-Element Film 

Shield

Multi-Element Film

TLD

Figure 6-1. Estimated dosimeter photon response characteristics.

6.3.3

Potential error in recorded dose is dependent on dosimetry technology response characteristics to 

each radiation type, energy, and geometry; the methodology used to calibrate the dosimetry system; 

and the extent of similarity between the radiation fields used for calibration and that present in the 

workplace. The potential error is much greater for dosimeters with significant variations in response, 

such as film dosimeters for low-energy photon radiation and the nuclear track emulsion and TLND for 

neutron radiation. 

Calibration

6.3.3.1 Beta/Photon Dosimeters

The beta/photon film dosimeters at PGDP were calibrated to 226Ra until 1980, when the calibration 

source was changed to 137Cs. The calibration to both 226Ra and 137Cs was free in air (no phantom) 

until the DOELAP procedures adopted in 1986 required phantoms. Hp(10) is defined with a phantom, 

in particular the ICRU slab phantom, which is a conservative practical definition of anterior–posterior 

whole-body dose to the standard ICRU spherical phantom (ICRU 1993). 

Introduction of on-phantom calibration of film dosimeters and replacement of 226Ra by 137Cs as the 

calibration source changed the relationship between recorded dose and Hp(10). In addition to

registration of the additional backscattered radiation, the generally lower energy photon spectrum 

from 226Ra in comparison with 137Cs (662 keV) gave a greater optical density for the same dose during 

calibration (Figure 6-1). In contrast, the effect of backscatter is to overestimate dose, and calibration 

with 226Ra tends to underestimate the dose relative to calibration with 137Cs. Because the photons at 

PGDP are of intermediate energies, some numerical dose adjustment should be considered for the 

early film dosimetry. The overall dose adjustment depends on these two factors, which act in 

opposite directions, as well as the spectrum of registered photons and the film dosimeter itself. 

In the 1980s, studies were carried out at a number of laboratories to assess changes from the 

on-phantom calibration mandated by the new DOELAP testing criteria (Fix et al. 1982; Wilson 1987;

Wilson et al. 1990; Taylor et al. 1995). While not exactly the same at all sites, most film dosimeters, 

like those at PGDP, had common features due to their evolution from the original work of Pardue, 

Goldstein, and Wollan (1944). The early badges were calibrated to exposure in free air. Laboratory 

tests at Hanford showed 8% and 4% increases, respectively, in dosimeter response for on-phantom 

exposures using 226Ra and 137Cs. With free-air calibration, the exposure to the wearer tends to be 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 11 of 25

overestimated by this amount, which is assumed to be similar for Paducah. Tests at Savannah River, 

on the other hand, indicated that film badge doses underestimated Hp(10) by 11.9% before 1986 and 

by 3.9% for 1986 (Taylor et al. 1995). Lacking site-specific data for PGDP, it is recommended that 

exposure-to-organ dose conversion factors in Appendix B of (NIOSH 2002) be employed for dose 

reconstruction at PGDP with no numerical adjustment to the recorded doses. This procedure is 

expected to be claimant-favorable. It allows for an overestimate of exposure, such as assessed in the 

Hanford studies, which should also be sufficient to offset effects due to the calibration source, if they 

are in the opposite direction.

For a number of years, ORNL used uranium beta as well as 226Ra gamma calibration curves to 

interpret film densities (Thornton, Davis, and Gupton 1961). The ratio of beta-to-gamma responses 

was tested in several ways. Films wrapped in a 7 mg/cm2 absorber were placed in contact with a slab 

of natural uranium. The densities per rad were found to be nearly the same as those produced from 

226Ra gamma rays measured behind a cadmium filter. In addition, stacks of film were exposed on a 

uranium surface, and the densities at various depths were used to extrapolate to the value for a depth 

of 7 mg/cm2

. This value was also nearly equal to that produced by the same dose from 226Ra photons 

behind the cadmium filter. Therefore, for beta radiation from natural uranium, the density produced 

per rad in film was equal to the density produced per rad behind the cadmium filter by 226Ra gamma 

rays. Analysts concluded that, for routine personnel dosimetry, film was equally sensitive for beta and 

gamma radiations. Because the film badge had a minimum absorber thickness of 80 mg/cm2

between the film and the source, the effective beta energy is needed to interpret the film density in 

terms of Hp(0.07). The radiation was routinely treated as 1.7-MeV beta particles from uranium, which 

are about 40% absorbed in 80 mg/cm2 (Thornton, Davis, and Gupton 1961). The determination of 

beta dose was thus specific to uranium.

6.3.3.2 Neutron Dosimeters

Calibration of neutron dosimeters for use at PGDP was appropriate for the work locations in which 

these dosimeters were worn. Dosimeter response was characterized in a manner that would 

represent the workplace (Martin Marietta 1994). Reference dosimetry for these measurements was 

evaluated with tissue-equivalent proportional counters (TEPCs). TEPCs provide an absolute measure 

of absorbed dose in a tissue-like material and, with an appropriate algorithm, an estimate of the 

neutron quality factor (PNL 1995). The basis for the calibration factor was developed using data 

obtained at the Y-12 plant in a room used to store an array of small canisters of UF4. Measurements 

were made with Bonner spheres at the same location. The average quality factor was 11, and the 

average energy range was 0.6 to 1.4 MeV (PNL 1990). 

In 1993, field measurements were made by ORNL representatives at the end row of the K-25 cylinder 

yard with a TEPC and a phantom with TLDs about 4 feet from the outside of a cylinder at about the 

middle of its length. The results were evaluated qualitatively because the dose rate was very low and 

an appropriate power supply was not available. The correction factors were similar to those in the 

Y-12 UF4 storage area and confirmed the appropriateness of these values. These correction factors 

apply to the PGDP TLNDs.

6.3.4

6.3.4.1 Beta/Photon Fields

Workplace Radiation Fields

PGDP operations are characterized by the relatively low-level external beta and photon radiation 

fields associated with uranium in feed materials, products, wastes, and contaminated equipment and 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 12 of 25

systems. Processed RU was present with natural, depleted, and enriched (up to 2% 235U by weight) 

abundances. (Section 6.3.4.3 describes potential sources for neutron exposure.) 

Table 6-2 summarizes the major sources of external radiation throughout PGDP operations (PACE 

and University of Utah 2000). The photon energy range of principal interest is 30 keV to 250 keV. 

Handling uranium material of these types did not, in general, produce areas with significantly elevated 

photon radiation. 

Table 6-2. Major radiation sources.

Nuclide Source Half-life

Energies (MeV) and abundances 

of major radiations

Alpha Beta (max) Gamma

U-238 Primary U isotope 4.51E9 yr 4.15 (21%)

4.20 (79%)

U-235 Primary U isotope 7.1E8 yr 4.21 (6%) 0.144 (11%)

4.37 (17%) 0.163 (5%)

4.40 (55%) 0.186 (57%)

4.60 (5%) 0.205 (5%)

U-234 Primary U isotope 2.47E5 yr 4.72 (28%) 0.053(0.12%)

4.77 (72%)

Th-234 Decay product 24.1 day 0.013 (9.8%)

0.103 (21%) 0.063 (3.5%)

0.193 (79%) 0.092 (3%)

0.093 (4%)

Pa-234m Decay product 1.17 min 2.29 (98%) 0.765 (0.3%)

1.001 (0.60%)

Th-231 Decay product 25.5 hr 0.206 (13%)

0.287 (12%) 0.026 (2%)

0.288 (37%) 0.084 (10%)

0.305 (35%)

Tc-99 Impurities from RU 2.12E5 yr 0.294 (100%) None

The major facilities and associated activities at PGDP are (BJC 2000):

• C-331, C-333, C-335, and C-337 – Gaseous Diffusion Process Buildings

• C-410/420 – UF6 Feed Plant

• C-310 – Purge and Product Withdrawal Building

• C-315 – Surge and Tails Withdrawal Building

• C-340 – Metals Plant

• C-400 – Decontamination and Cleaning Building

• C-720 – Maintenance Building

The buildings with the greatest potential for elevated direct radiation levels were C-340, C-410, C-420, 

and the cascade buildings (PACE and University of Utah 2000). From 1952 to approximately 1980, 

the major sites of potential exposure to radioactive material were buildings involved in the conversion 

of UO3 powder to enriched UF6 in solid or gaseous form, UF4 and uranium metals recovery operations, 

and the decontamination building. Feed and enrichment operations were in Buildings C-410, C-420, 

C-331, C-333, C-335, C-337, C-310, and C-315, while UF4 recovery and uranium recovery were in 

Building C-340. The decontamination operation was in Building C-400. The oxide conversion 

building, C-420, was where UO3 powder (clean or recycled) was received and converted to UF4. 

From Building C-420, material went to Building C-410, the feed plant, for conversion to UF6. Last, UF6

was processed through the cascade buildings (C-331, C-333, C-335, and C-337). Enriched UF6 was 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 13 of 25

withdrawn in Building C-310, the product withdrawal building, while depleted UF6 was removed in 

Building C-315, the tails withdrawal building. Table 6-3 lists the principal buildings, sources for 

external dose, and periods of operation.

Table 6-3. Buildings and periods of operation.

Site facilities Source for external dose

Operation

Begin End

C-310 Purge and Product Withdrawal UF6 process equipment and cylinders 1953 1999

C-315 Surge and Tails Withdrawal UF6 process equipment and cylinders 1953 1999

C-331, C-333, C-335, C-337 Gaseous Diffusion 

Process Buildings

UF6 process equipment and cylinders 1953 1964

1969 1970

1972 1976

C-340 Reduction and Metals Facility Process equipment, contaminated 

floors

1957 1962

1967 1977

C-400 Decontamination and Cleaning Buildings UF6 process equipment and cylinders 1952 1990

C-410 UF6 Feed Plant and C-420 Oxide 

Conversion Plant

Process equipment, contaminated 

floors

1953 1964

1968 1977

C-415 Feed Plant Storage Building Radioactive source storage area 1953 1977

C-745 A-V Cylinder Yards UF6 cylinders 1953 (estimated) Ongoing

PGDP also processed RU. The feed material contained trace amounts of radioactive impurities not 

present in natural uranium feed material. Because these impurities were present in such minute 

concentrations, their radiological impact was usually negligible. However, some routine chemical 

processes would concentrate them. From an external dose standpoint, the most significant impurity 

found in RU is the pure beta emitter, 99Tc, which tends to deposit in enrichment equipment and pocket

in the higher sections of the diffusion cascade (DOE 2000). Technetium-99 was also concentrated for 

recovery and removal. The relatively low-energy beta particles (maximum 294 keV) from 99Tc pose 

minimal external exposure potential because of their limited range. Neither film nor TLD efficiently 

detect them, particularly in the presence of uranium. Clothing and gloves provide adequate shielding. 

Skin contamination is the only credible scenario where significant shallow dose could occur from 99Tc. 

Table 6-4 shows the principal locations where and periods during which the recovery operations at 

PGDP are believed to have taken place (PACE AND UNIVERSITY OF UTAH 2000).

Table 6-4. Technetium-99 recovery 

operations.

Building Began Terminated

C-710 Before 1959 ~1959

C-400 ~1959 ~1975

6.3.4.2 Workplace Beta/Photon Dosimeter Response

Essentially all PGDP radiological work areas involved photon and beta radiation characteristic of 

operations involving uranium at low enrichments. As discussed in Section 6.3.3.1, the recorded 

responses of the PGDP beta/photon film dosimeters are claimant-favorable and need no adjustment. 

6.3.4.3 Neutron Fields

While neutrons occur in some areas at PGDP, the measured levels are low. There are no locations 

identified where measurable neutron dose was encountered (Martin Marietta 1994). Several studies 

have evaluated neutron fields at gaseous diffusion plants (PNL 1995; Cardarelli 1996); these studies 

confirm Martin Marietta (1994). Cylinder yards, feed and withdraw areas, and locations where 

uranium forms deposits in the cascade have been investigated (Cardarelli 1996). These studies 

identified the storage cylinders, which contained either depleted UF6 (tails) or enriched UF6 (product), 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 14 of 25

as areas where neutron fields could represent an exposure hazard. Estimates of dose equivalent 

rates range from 0.007 to 0.34 mrem/hr; associated quality factors range from 7 to 10. A 

representative average value is 0.2 mrem/hr based on a quality factor of about 10 (PNL 1995; 

Cardarelli 1996). Estimates of average neutron energies ranged from 0.25 to 0.56 MeV (PNL 1995). 

Neutron monitoring of individuals was performed during a UF6 cylinder-painting project (Meiners 

1999). Results of this project indicated a neutron-to-photon dose equivalent ratio of approximately 

1 to 5, based on a quality factor of 10. The associated neutron-to-photon absorbed dose ratio is 

1 to 50. 

6.3.4.4 Workplace Neutron Dosimeter Response

Quantitative monitoring for neutron dose began at PGDP in 1998. TLNDs were used in conjunction 

with appropriate work field calibration factors. Before 1998, the beta/photon badge assembly 

contained a neutron-sensitive element (NTA, NTB, Eastman Kodak Type 2). This element was 

processed only when requested. (NTA film had an energy threshold of about 0.5 MeV.) A review of 

data does not indicate the assignment of neutron dose before 1998. 

6.4 ADJUSTMENTS TO RECORDED DOSE

6.4.1

The recorded doses varied in reporting units depending on regulatory requirements and dose 

definitions (both national and international). The current reporting unit used by DOE is the millirem, a 

unit of dose equivalent. The international unit of dose equivalent is the milliseivert, which is equivalent 

to 100 mrem. Since 1986, deep dose equivalents at PGDP have been based on DOELAP calibration 

to Hp(10) and require no adjustment. Before 1986, TLDs were calibrated in air to 137Cs, which is 

nearly equivalent to an Hp(10) on-phantom 137Cs calibration. No adjustment to the measured TLD 

penetrating photon dose is necessary. As discussed in Section 6.3.3.1, the earlier film badge deep 

doses are claimant-favorable and require no numerical adjustment.

Photon Dose

6.4.2

The early film dosimeters were calibrated to uranium for nonpenetrating radiation. No numerical 

adjustment of recorded shallow doses is recommended. Incident reports are a possible source that 

can be consulted for investigations of nonroutine beta exposures and dose assessment.

Nonpenetrating Dose

6.4.3

The measured neutron energies at PGDP are between 0.10 MeV and 2.0 MeV, for which the 

International Commission on Radiological Protection Publication 60 (ICRP 1990) radiation weighting 

factor is 20. Therefore, the reported neutron dose equivalent should be multiplied by a factor of 2 to 

be consistent with the ICRP (1990) recommendations to be used for reconstruction (NIOSH 2002). 

This factor should be applied to both measured and missed neutron doses.

Neutron Dose

6.5 MISSED DOSE

Missed deep and shallow doses have been examined for three groups of PGDP workers as follows: 

1. A zero dose was recorded but the worker was not monitored (majority of workers from 1953 to 

July 1960).
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 15 of 25

2. A zero dose was recorded for the dosimeter system for any response less than the MDL.

3. There was no recorded dose because workers were not monitored, or the dosimetry record is 

not available. 

Neutron dose rates at PGDP were low (Martin Marietta 1994). Neutron dosimeters were not routinely 

assigned and doses recorded until about 1998. Neutron doses reported before 1998 were based on a 

conservative calibration associated with a neutron-sensitive element incorporated in the beta/gamma 

dosimeter. Application of a neutron-to-gamma dose equivalent ratio of 1 to 5 appears to be a 

satisfactory, claimant-favorable option, because the photon dose is reliably measured. This ratio can 

be applied to selected work activities.

6.5.1

Methods to be considered when there is no recorded dose for a period during a working career have 

been examined by Watson et al. (1994). In general, estimates of unmonitored dose can be made by 

using dose results for coworkers or the recorded dose before and after the period when they were not 

monitored. However, these situations require careful examination. The dose reconstructor should 

consider all reasonable methods and assign the most appropriate dose based on the employees’ job 

description and work locations. NIOSH (2002) cites several different models.

Estimating Missed and Unmonitored Photon Deep Dose

For Group 2, the missed dose for dosimeter results that are less than the MDL is particularly important 

for earlier years, when MDLs were higher and dosimeter exchange was more frequent. NIOSH 

(2002) describes an acceptable, claimant-favorable estimate of the maximum potential missed dose 

as one-half the MDL multiplied by the number of zero dose results (the MDL/2 method). The last 

column in Table 6-1 lists the resultant estimates of the annual missed dose for Group 2 for different 

years at PGDP. 

If it is definitely established that the employee was not a radiation worker, then the unmonitored deep 

dose for that period can be assigned as the on-site ambient dose.

Otherwise, an individual in Group 1 or 3 should be treated as a radiation worker. The unmonitored 

deep dose can then be approached in two ways. First, the same assignment of missed dose as for 

Group 2, from the last column of Table 6-1, can be considered. However, for the period 1953 through 

July 1960, with the frequent (weekly) dosimeter exchange and relatively large MDL, the resulting 

implied annual missed dose of 1 rem is probably unrealistically large for many unmonitored persons in 

both Groups 1 and 3. Figure 6-2 shows the distribution of individual annual deep dose equivalent for 

monitored workers for the years 1953 to 1974 (Baker c. 1995). Few of these individuals received as 

much as 1 rem in any given year.

A second, alternative, approach for Group 1 or 3 is to base the unmonitored dose estimate on 

exposure data compiled in (PACE and University of Utah 2000) for monitored PGDP workers. The 

first four columns in Table 6-5 (taken from Table 7.4 of the PACE Report) show the number of 

monitored workers, their average recorded deep dose, and the maximum individual deep dose for 

each year from 1953 through 1988 (zero doses were not included.) For use in the present dose 

reconstruction, it was assumed that the exposure data for each year could be represented by a 

lognormal distribution with a geometric mean (GM) equal to the average shown in column 3 of Table 

6-5 and a 99th percentile equal to the maximum in column 4. With these assumptions, the geometric 

standard deviation (GSD) of the lognormal distribution, shown in the last column, was computed. The 

two parameters, GM and GSD, thus determine the dose distribution assumed for the monitored 

workers for each year. The values from columns 3 and 5 in Table 6-5 can be entered directly into the 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 16 of 25

Interactive RadioEpidemiological Program (IREP). Unmonitored dose values obtained both from the 

use of Table 6-5 and from Table 6-1 should be considered. Knowledge of specific conditions known 

for some workers at some periods could have a bearing on which value should be considered more 

appropriate. In the absence of such information, the larger of the two values should be used as 

claimant-favorable.

Figure 6-2. Historical distribution of deep dose equivalent (Baker c. 1995).

6.5.2

The procedure for assessing missed and unmonitored shallow dose is similar to that for the missed 

deep dose.

Estimating Missed and Unmonitored Shallow Dose

For Group 2, the missed annual shallow dose equivalent is given in the last column of Table 6-1, in 

keeping with the MDL/2 method of evaluation. Figure 6-3 shows the historical data for the distribution 

of shallow dose equivalent for monitored workers at PGDP (Baker c.1995). When compared with 

Figure 6-3, this assessment of annual missed shallow dose for Group 2 is seen to be claimantfavorable.

For non-radiological workers identified in Groups 1 and 3, the unmonitored shallow dose can be 

assigned as the environmental dose. Other individuals in these groups should be regarded as 

radiation workers, for whom the same estimate as that used for Group 2 should be considered. 

Alternatively, Table 6-6 can be used, which is based on the shallow-dose data for monitored workers 

taken from Baker (c. 1995) and shown in Figure 6-3. It was assumed that the dose distribution for 

each year could be represented by a lognormal function with GM equal to the average shown in 

column 2 of Table 6-6 and a 95th percentile equal to the maximum in column 3. The resultant

calculated GSD is shown in the last column. The GM and GSD can be entered into IREP. Values for 

unmonitored shallow dose obtained for Groups 1 and 3 from Table 6-6 should be compared with 

those determined by the MDL/2 method from Table 6-1. Knowledge of specific job conditions and 

location should be considered in judging which of the two estimates is more appropriate. In the 

absence of such information, the larger estimate should be assigned as claimant-favorable.
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Table 6-5. Average recorded deep dose and maximum for any 

single worker by year (PACE and University of Utah 2000).a

 

Year

Number of 

workers

Average dose, 

GM (rem)

Maximum

dose (rem) GSD (rem)

1953 223 0.1398 0.820 2.14

1954 284 0.2835 1.580 2.09

1955 417 0.2419 2.500 2.72

1956 471 0.3586 4.700 3.02

1957 669 0.2517 3.190 2.97

1958 661 0.1853 3.630 3.59

1959 570 0.2015 2.360 2.88

1960 526 0.2011 2.510 2.95

1961 1,690 0.1770 2.530 3.13

1962 1,479 0.1495 2.980 3.61

1963 1,311 0.1441 3.040 3.70

1964 1,289 0.0734 1.860 4.00

1965 1,128 0.0341 1.610 5.23

1966 1,138 0.0371 1.470 5.19

1967 1,143 0.0498 1.120 3.80

1968 1,241 0.0618 1.400 3.82

1969 1,270 0.0733 1.970 4.11

1970 1,273 0.0417 0.840 3.63

1971 1,254 0.0624 1.380 3.78

1972 1,288 0.0589 1.760 4.30

1973 1,404 0.0530 1.830 4.57

1974 1,624 0.0265 1.030 4.81

1975 2,013 0.0501 1.049 3.69

1976 2,426 0.0351 1.224 4.59

1977 2,643 0.0232 0.742 4.42

1978 2,613 0.0399 0.359 2.57

1979 2,487 0.0082 0.364 5.09

1980 2,308 0.0182 0.344 3.53

1981 1,840 0.0076 0.420 5.60

1982 1,617 0.0065 0.350 5.53

1983 1,452 0.0067 0.340 5.39

1984 1,434 0.0092 0.420 5.15

1985 1,365 0.0061 0.350 5.69

1986 1,244 0.0096 0.490 5.41

1987 1,275 0.0080 0.470 5.74

1988 1,359 0.0065 0.720 7.54

a. As explained in text, columns 3 and 5, respectively, show the GM and 

GSD of lognormal distribution used to describe the data.

Significant nonroutine beta doses, such as could occur from skin contamination events, could be

addressed in specific incidence reports. In such cases, assessments based on investigations 

conducted at the time of the incident should be considered the best resource for dose reconstruction.

Potential doses from 99Tc skin contamination have been evaluated by using the VARSKIN computer 

code. The calculated shallow dose rate from uniform 99Tc skin contamination is 0.0016 mrem/hr per 

dpm/cm2 (Swinth 2004). Technetium-99 has proven to be difficult to remove from skin. Therefore, the 

integrated shallow dose resulting from 99Tc skin contamination could be relatively large. For example, 

with a residence half-time of 1.5 days, the dose is 0.081 mrem per dpm/cm2 of initial contamination.
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 18 of 25

Figure 6-3. Historical distribution of shallow dose equivalent (Baker c. 1995).

Table 6-6. Average recorded shallow dose and 

maximum for any worker by year (Baker c.1995).

Year

Average dose, 

GM (rem)

Maximum

dose (rem) GSD (rem)

1953 0.539 4 2.36

1954 0.677 7 2.73

1955 0.776 9 2.86

1956 0.853 12 3.11

1957 0.834 11 3.03

1958 0.809 11 3.07

1959 0.783 10 2.98

1960 0.699 10 3.13

1961 0.734 8 2.79

1962 0.719 10 3.09

1963 0.645 8 2.95

1964 0.547 4 2.35

1965 0.511 2 1.80

1966 0.511 3 2.14

1967 0.528 6 2.84

1968 0.563 8 3.12

1969 0.616 5 2.46

1970 0.552 3 2.07

1971 0.631 7 2.81

1972 0.640 10 3.25

1973 0.679 10 3.17

1974 0.578 7 2.92

In general, direct external beta dose from 99Tc is minimal. The unshielded shallow dose rate to bare 

skin (no clothing) at a distance of 10 cm in air from a uniformly contaminated surface is about 

1 × 10-4 mrem/hr per dpm/cm2

, as estimated with VARSKIN. The dose rate at 30 cm is only about 

1 × 10-6 mrem/hr per dpm/cm2

. Table 6-7 summarizes these three benchmark values for shallow 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 19 of 25

dose equivalent rate as determined from VARSKIN for skin contamination and for external exposure 

with intervening air.

Table 6-7. Shallow dose equivalent rates for 

99Tc.

Condition

Dose-equivalent rate 

(mrem/hr per dpm/cm2)

Skin contamination 1.6 × 10-3

External, 10 cm air 1.0 × 10-4

External, 30 cm air 1.0 × 10-6

It is possible that some skin contamination events involving 99Tc occurred without being detected at 

the time. In some cases, therefore, it could be appropriate to consider an additional skin dose 

component for a reported shallow dose of a worker who could have had direct contact with 99Tc. In 

the absence of specific data, the dose reconstructor must make assumptions about the number of 

times per year that an affected skin region could have been contaminated and the extent of each 

contamination. For example, the reconstructor could assume a monthly contamination event at a 

specific location on the skin with an average level of 25,000 dpm/100 cm2 (the action limit for 99Tc 

contamination on work surfaces and hand tools at PGDP). With the assumed residence half-time of 

1.5 days, the annual shallow dose equivalent would be 240 mrem (12 × 250 dpm/cm2 × 0.081 mrem 

per dpm/cm2

). The direct external dose rate at a distance of 10 cm from a surface contaminated at 

this same level would be 0.025 mrem/hr (250 dpm/cm2 × 10-4 mrem/h per dpm/cm2

). At 30 cm, the 

rate would be 0.00025 mrem/hr.

6.5.3

A neutron component should be added to the annual dose of individuals who worked in the cylinder 

yard before 1998. However, careful consideration should be given to work history. In general, only 

workers who were near cylinders for extended periods have the potential for neutron exposure. 

Estimates should be based on the neutron-to-photon ratio of 1 to 5 for dose equivalent, as determined 

from the survey conducted at PGDP (Meiners 1999). The neutron dose equivalent should then be 

multiplied by the ICRP (1990) factor of 2. 

Estimating Missed Neutron Dose

6.6 UNCERTAINTY

PGDP has historically used ORNL personnel dosimeter services. ORNL has assessed the standard 

error in the recorded film-badge dose as ±30% for photons of all energies (ORAU 2004). The 

standard error for beta dose is the same (or somewhat larger for unknown mixtures of beta/gamma 

dose). Thus, the film-badge dose uncertainty is 1.3. The uncertainty in the TLD dose is 1.15 (ORAU 

2004), which is consistent with NIOSH (2002).

6.7 DOSE RECONSTRUCTION

As much as possible, dose to individuals should be based on dosimetry records. It is important to 

distinguish between the recorded nonpenetrating and penetrating doses and the actual Hp(0.07) and 

Hp(10). The following list summarizes appropriate information:

• Dosimetry records that provide nonzero beta-photon values for Hp(10) and Hp(0.07) are 

considered adequate. No numerical adjustment of the doses is required. Beta energies are

greater than 15 keV and photon energies should be considered to be in the range 30 keV to 

250 keV.
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 20 of 25

• Workers for whom dosimetry records provide zero beta-photon values for Hp(10) and Hp(0.07)

should have missed dose assigned on the basis of MDL/2 times the number of zero results, as 

described in Sections 6.5.1 and 6.5.2 (NIOSH 2002). 

• Individuals with no dose recorded might or might not have been radiological workers. If it is 

definitely established that the individual was not a radiation worker, then the assigned missed 

dose is the environmental dose discussed in the Occupational Environmental Dose portion of 

this PGDP Site Profile. Otherwise, the missed dose is to be estimated as described in Section 

6.5. No numerical adjustments to the missed dose are necessary.

• Reported and missed neutron dose equivalents should be multiplied by 2 to adjust for ICRP 

(1990).

• Cylinder yard workers for whom no neutron dose is recorded should have missed neutron 

dose equivalent estimate assigned based on a neutron-to-photon ratio of 1 to 5 for dose 

equivalent (Meiners 1999). Multiply the estimated neutron dose equivalent by 2 to adjust for 

ICRP (1990).

• Special attention should be paid to the possibility of skin contamination incidents for workers 

involved with 99Tc recovery operations (Section 6.5.2).

• Uncertainty is discussed in Section 6.6.

6.8 ORGAN DOSE

NIOSH (2002) discusses the conversion of measured doses to organ dose equivalent, and Appendix 

B of that document contains the appropriate dose conversion factors for each organ, radiation type, 

and energy range based on the type of monitoring performed. In some cases, simplifying 

assumptions are appropriate. 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 21 of 25

REFERENCES

BJC (Bechtel Jacobs Company, LLC), 2000, Recycled Uranium Mass Balance Project, Paducah 

Gaseous Diffusion Plant Site Report, BJC/PGDP-167, Paducah, Kentucky.

Baker, R. C., ca. 1995, Occupational Radiation Exposure Experience, Paducah Gaseous Diffusion 

Plant, Union Carbide Corporation, Nuclear Division, Paducah, Kentucky.

Cardarelli J. J., 1996, Health Hazard Evaluation Report, 96-0198-2651, Portsmouth Gaseous 

Diffusion Plant, Piketon, Ohio.

DOE (U.S. Department of Energy), 1986, Department of Energy Standard for the Performance 

Testing of Personnel Dosimetry Systems, DOE/EH-0027, Washington, D.C.

DOE (U.S. Department of Energy), 2000, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136-2000, Washington, D.C.

Fix, J. J., J. M. Hobbs, P. L. Roberson, D. C. Haggard, K. L. Holbrook, M. R. Thorson, and F. M. 

Cummings, 1982, Hanford Personnel Dosimeter Supporting Studies FY-1981, PNL-3736, 

Pacific Northwest Laboratory, Richland, Washington.

Fix, J. J., L. Salmon, G. Cowper, and E. Cardis, 1997, “A Retrospective Evaluation of the Dosimetry 

Employed in an International Combined Epidemiologic Study,” Radiation Protection Dosimetry,

volume 74, pp. 39-53.

ICRP (International Commission on Radiological Protection), 1990, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

ICRU (International Commission on Radiation Units and Measurements), 1993, Conversion 

Coefficients for Use in Radiological Protection Against External Radiation, Report 57, 

Bethesda, Maryland.

Martin Marietta (Martin Marietta Energy Systems, Inc.), 1994, Centralized External Dosimetry System 

(CEDS), Technical Basis for the Centralized Dosimetry System, Oak Ridge, Tennessee.

Meiners, S., 1999, Paducah UF6 Cylinder Painting Project, Bechtel Jacobs Company, LLC, Paducah, 

Kentucky.

NIOSH (National Institute for Occupational Safety and Health), 1993, “Epidemiologic Use of 

Nondetectable Values in Radiation Exposure Measurements,” NIOSH Research Issues 

Workshop, September 9-10, 1993, Cincinnati, Ohio.

NIOSH (National Institute for Occupational Safety and Health), 2002, External Dose Reconstruction 

Implementation Guideline, Revision 1, OCAS-IG-001, Office of Compensation Analysis and 

Support, Cincinnati, Ohio.

NRC (National Research Council), 1989, Film Badge Dosimetry in Atmospheric Nuclear Tests, 

National Academy Press, Washington, D.C.
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ORAU (Oak Ridge Associated Universities), 2004, Technical Basis Document for ORNL Site –

Occupational External Dose, ORAUT-TKBS-0012-6, Oak Ridge, Tennessee.

PACE (Paper, Allied Industrial, Chemical and Energy Workers International Union) and University of 

Utah, 2000, Exposure Assessment Project at the Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky.

Pardue, L. A., N. Goldstein, and E. O. Wollan, 1944, Photographic Film as a Pocket Radiation 

Dosimeter, CH-1553-A-2223, University of Chicago, Metallurgical Laboratory, Chicago, Illinois. 

PNL (Pacific Northwest Laboratories), 1990, Neutron Dose Equivalent and Energy Spectra 

Measurements at the Oak Ridge National Laboratory and Y-12 Plant, Richland, Washington. 

(Reference not available; summary of information provided by Kim McMahan)

PNL (Pacific Northwest Laboratories), 1995, Enriched Uranium Cylinder Neutron Characterization at 

the Portsmouth Gaseous Diffusion Plant, Richland, Washington.

Swinth, K. L., 2004,” electronic mail correspondence, March 5 and 19, 2004. 

Taylor, G. A., K. W. Crase, Thomas R. LaBone, and W. H. Wilkie, 1995, A History of Personnel 

Radiation Dosimetry at the Savannah River Site, WSRC-RP-95-234, Westinghouse Savannah 

River Company, Aiken, South Carolina.

Thierry-Chef, I., F. Pernicka, M. Marshall, E. Cardis, and P. Andreo, 2002, “Study of a Selection of 10 

Historical Types of Dosimeter: Variation of the Response to Hp(10) with Photon Energy and 

Geometry of Exposure,” Radiation Protection Dosimetry, volume 102, pp. 101-113.

Thornton, W. T., D. M. Davis, and E. D. Gupton, 1961, The ORNL Badge Dosimeter and its Personnel 

Monitoring Applications, ORNL-3126, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Watson, Jr., J. E., J. L. Wood, W. G. Tankersley, and C. M. West, 1994, “Estimation of Radiation 

Doses for Workers Without Monitoring Data for Retrospective Epidemiologic Studies,” Health 

Physics, volume 67, number 4, pp. 402-405.

Wilson, R. H., 1987, Historical Review of Personnel Dosimetry Development and its Use in Radiation 

Protection Programs at Hanford: 1944 Through 1989, PNL-6125, Pacific Northwest 

Laboratory, Richland, Washington.

Wilson, R. H., J. J. Fix, W. V. Baumgartner, and L. L. Nichols, 1990, Description and Evaluation of the 

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Northwest Laboratory, Richland, Washington.
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 23 of 25

GLOSSARY

absorbed dose 

Radiation energy deposited in tissue or other material divided by the mass of the tissue or 

material.

albedo dosimeter

A dosimeter that measures slow neutrons generated by higher energy neutrons incident on the 

body and that reflect back into the dosimeter. 

beta radiation

Radiation consisting of electrons emitted spontaneously from the nuclei of certain radioactive 

elements.

curie

A special unit of activity. One curie exactly equals 3.7 × 1010 nuclear transitions per second.

deep dose equivalent (Hd)

The dose equivalent at the depth of 10 mm in tissue.

dose equivalent (H)

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The special unit is the rem. When D is expressed in gray, H is in sieverts (1 sievert = 100 

rem.)

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. (See film dosimeter, neutron 

film dosimeter, thermoluminescent dosimeter.)

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external and/or internal sources of radiation.

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, 

and/or extremities. This includes the fabrication, assignment, and processing of dosimeters as 

well as interpretation and documentation of the results.

film

In general, a "film packet" that contains one or more pieces of film in a light-tight wrapping. 

When developed, the film has an image caused by radiation that can be measured using an 

optical densitometer. 

film dosimeter

A small packet of film in a holder that attaches to a wearer.

gamma rays 

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Physically, gamma 
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 24 of 25

rays are identical to X-rays of high energy, the only essential difference being that X-rays do 

not originate in the nucleus.

gray (Gy)

The special name for the SI unit of absorbed dose (1 gray = 1 joule/kilogram).

kerma

Sum of initial kinetic energies of all charged particles (including Auger electrons) liberated by 

uncharged radiation per unit mass. Units are rad and gray. The word derives from kinetic 

energy released per unit mass.

minimum detection level

The minimum quantifiable exposure or neutron flux that can be detected by a dosimeter.

neutron

A basic nuclear particle that is electrically neutral, having nearly the same mass as the 

hydrogen atom.

neutron film dosimeter

A film dosimeter that contains a nuclear track emulsion film packet, such as NTA, NTB, or 

Eastman Type 2.

nuclear emulsion

Often referred to as NTA film and used to measure personnel dose from neutron radiation.

nuclear track emulsion, type A (NTA)

A film that is sensitive to fast neutrons. The developed image has tracks caused primarily by 

recoil protons. Tracks can be seen by using an appropriate imaging capability such as oil 

immersion and a 1,000-power microscope or a projection capability.

personal dose equivalent Hp(d)

Represents the dose equivalent in soft tissue below a specified point on the body at an 

appropriate depth d. The depths selected for personnel dosimetry are 0.07 mm and 10 mm, 

respectively, for the skin and body. These are noted as Hp(0.07) and Hp(10), respectively.

photon

A unit, or particle, of electromagnetic radiation consisting of X- and/or gamma rays.

rad

The traditional unit of absorbed dose (1 rad = 100 ergs per gram of material absorbing the 

radiation energy; 1 rad = 0.01 gray).

radiation

Alpha, beta, neutron, and photon radiation.

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, and gamma rays 

from unstable nuclei.
Effective Date: 03/29/2005 Revision No. 01 Document No. ORAUT-TKBS-0019-6 Page 25 of 25

rem

The traditional unit of dose equivalent, which is equal to the product of the absorbed dose in 

rad and the quality factor of the radiation (1 rem = 0.01 sievert).

rep (roentgen-equivalent-physical)

A historical unit used to report beta exposures, usually recorded in millirep; equivalent to 93 

ergs of energy per gram and roughly the same as a rem. 

roentgen (R)

A unit of exposure to gamma (or X-ray) radiation. It is defined precisely as the quantity of 

gamma (or X-) rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 kilogram of dry 

air at standard temperature and pressure. An exposure of 1 R is approximately equivalent to 

an absorbed dose of 1 rad in soft tissue for higher (≥100 keV) energy photons.

recycled uranium (RU)

Uranium recovered from used reactor fuel; isotopic activity ratios listed as:

Isotope Activity fraction

234U 0.8489

235U 0.0120

236U 0.1388

238U 0.0003 or 0.0004 (both listed)

shallow absorbed dose (Ds)

The absorbed dose at a depth of 0.07 mm in a material of specified geometry and 

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 mm in tissue.

sievert (Sv)

The SI unit for dose equivalent (1 sievert = 100 rem).

skin dose

Absorbed dose at a tissue depth of 7 mg/cm2 at about 0.07 mm depth in tissue.

thermoluminescence

Phosphorescence developed by heating a previously excited material.

thermoluminescent dosimeter (TLD)

A device used to measure radiation dose. It consists of a holder containing solid chips of 

material that when heated will release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose.

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1000 mg/cm2

); however, 

this term is also used to refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin or a radiograph.